1. Field of the Invention
This invention relates to a method for measuring the subcritical neutron multiplication factor, Keff of a nuclear reactor, and more particularly, to a method for determining all reactivity changes that occur while a core of a nuclear reactor is subcritical.
2. Related Art
In a pressurized water reactor power generating system, heat is generated within the core of a pressure vessel by a fission chain reaction occurring in a plurality of fuel rods supported within the core. The fuel rods are maintained in space relationship within fuel assemblies with the space between fuel rods forming coolant channels through which borated water flows. Hydrogen within the coolant water moderates the neutrons emitted from enriched uranium within the fuel to increase the number of nuclear reactions and thus increase the efficiency of the process. Control rod guide thimbles are interspersed within the fuel assemblies in place of fuel rod locations and serve to guide control rods which are operable to be inserted or withdrawn from the core. When inserted, the control rods absorb neutrons and thus reduce the number of nuclear reactions and the amount of heat generated within the core. Coolant flows through the assemblies out of the reactor to the tube side of steam generators where heat is transferred to water in the shell side of the steam generators at a lower pressure, which results in the generation of steam used to drive a turbine. The coolant exiting the tube side of the steam generator is driven by a main coolant pump back to the reactor in a closed loop cycle to renew the process.
The power level of a nuclear reactor is generally divided into three ranges: the source or startup range, the intermediate range, and the power range. The power level of the reactor is continuously monitored to assure safe operation. Such monitoring is typically conducted by means of neutron detectors placed outside and inside the reactor core for measuring the neutron flux of the reactor. Since the neutron flux in the reactor at any point is proportional to the fission rate, the neutron flux is also proportional to the power level.
Fission and ionization chambers have been used to measure flux in the source, intermediate and power range of a reactor. Typical fission and ionization chambers are capable of operating at all normal power levels, however, they are generally not sensitive enough to accurately detect low level neutron flux emitted in the source range. Thus, separate low level source range detectors are typically used to monitor neutron flux when the power level of the reactor is in the source range.
The fission reactions within the core occur when free neutrons at the proper energy level strike the atoms of the fissionable material contained within the fuel rods. The reactions result in the release of a large amount of heat energy which is extracted from the core in the reactor coolant and in the release of additional free neutrons which are available to produce more fission reactions. Some of these released neutrons escape the core or are absorbed by neutron absorbers, e.g., control rods, and therefore do not cause traditional fission reactions. By controlling the amount of neutron absorbent material present in the core, the rate of fission can be controlled. There are always random fission reactions occurring in the fissionable material, but when the core is shut down, the released neutrons are absorbed at such a high rate that a sustained series of reactions do not occur. By reducing the neutron absorbent material until the number of neutrons in a given generation equals the number neutrons in the previous generation, the process becomes a self sustaining chain reaction and the reactor is said to be in “critical”. When the reactor is critical, the neutron flux is six or so orders of magnitude higher than when the reactor is shut down. In some reactors, in order to accelerate the increase in neutron flux in the shut down core to achieve practical transition intervals, an artificial neutron source is implanted in the reactor core among the fuel rods containing the fissionable material. This artificial neutron source creates a localized increase in the neutron flux to aid in bringing the reactor up to power.
In the absence of a neutron source, the ratio of the number of free neutrons in one generation to those in the previous generation is referred to as the “neutron multiplication factor” (Keff) and is used as a measure of the reactivity of the reactor. In other words, the measure of criticality for a nuclear core is Keff, that is, the ratio of neutron production to total neutron loss contributable to both destruction and loss. When Keff is greater than 1, more neutrons are being produced than are being destroyed. Similarly, when Keff is less than one, more neutrons are being destroyed than are being produced. When Keff is less than one, the reactor is referred to as being “subcritical”. Until relatively recently, there has been no direct method for measuring when criticality will occur from the source range excore detectors. Plant operators typically estimated when criticality will occur through a number of methods. One method for estimating when criticality will occur is made by plotting the inverse ratio of the count rate obtained from the source range detector as a function of the change in conditions being used to bring the plant critical, e.g., withdrawal of the control rods. When the plant goes critical, the source range count rate approaches infinity and hence, the Inverse Count Rate Ratio (ICRR) goes to zero. Due to the physics of the reaction occurring within the core of the reactor, the ICRR curve is almost always convex, and sometimes concave. Therefore, estimating the conditions under which the plant will go critical from the ICRR curve is subject to much uncertainty, but also subject to considerable scrutiny by the Nuclear Regulatory Commission and Institute of Nuclear Power Operations.
More recently, a method has been devised for directly predicting when the reactor will go critical. The method is described in U.S. Pat. No. 6,801,593. In accordance with the method, the reactivity of the reactive core is increased while monitoring an output of a source range detector. The inverse count rate ratio from the output of the detector is determined periodically during a transient portion of the output. A correction factor is applied to the inverse count rate ratio data and the data is plotted as function of time. The correction factor linearizes the inverse count rate ratio so that the curve can be predictably extrapolated. The method thus describes a spatially corrected inverse count rate core reactivity measurement process. However, this method does not address the accuracy of the core reactivity measurement, which is dependent on the accuracy of the measured neutron radiation levels. In particular, it is very important that fractional changes in the measured neutron levels are determined accurately. The largest neutron measurement error component in a properly operating neutron radiation detector is typically caused by what is commonly called a “background signal”. The background signal induces a response in the detector measurement that is not caused by source neutrons. This results in errors in the measured core reactivity changes. In order to improve the accuracy of the neutron population measurement, and obtain a corresponding improvement in accuracy in the inverse count rate ratio reactivity measurement process, it is necessary to remove the background signal component from the measurement before the measurement is used to calculate the reactivity change. To this point in time, there has been no direct method of determining the background signal content in a neutron signal measurement from the typical neutron detectors used in commercial nuclear power facilities.
Accordingly, a method is desired that can be demonstrated to produce an accurate determination of the background content of neutron signals measured at commercial nuclear power facilities. Furthermore, such a method is desired that will not require any changes to existing commercial nuclear power plant equipment or operating practices.